DLC-0031 ZZ-FEWG1, ZZ-GEDT1/4

ZZ-FEWG1, ZZ-GEDT1/400, ZZ-GPDT1/402B, ZZ-NEDT1/401.

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FEWG1/H

FORMAT: ANISN

NUMBER OF GROUPS: 37-neutron, 21-gamma. Energy range 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays.

NUCLIDES: H, 3H, 6Li, 7Li, 9Be, 10B, 11B, 12C, N, O, F, Na, Mg, Al, Si, 31P, 32S, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, 181Ta, 182W, 183W, 184W, 186W,Pb, 235U, 238U, 238Pu, 240Pu.

ORIGIN: DNA Working Cross Section Library and the Evaluated Nuclear Data File (ENDF/B)

WEIGHTING SPECTRUM: 1/E and 300K Maxwellian weighting spectra were used for neutrons.

GEDT1/400

FORMAT: Gamma-ray cross section library in SAM-CE format.

NUMBER OF GROUPS: Data are given for the energy range from 1 keV to 100 MeV.

NUCLIDES: Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occurring elements.

ORIGIN: DLC-0007

WEIGHTING SPECTRUM: --

GPDT1/402B

FORMAT: Gamma-ray production data in SAM-CE format.

NUMBER OF GROUPS: --

NUCLIDES: Data are included for 1H, N, O, Al, Si, Ca, Pb, 7Li, 10B, 12C, Mg, Fe, 235U, 238U, Cl, 182W, 183W, 184W, 186W.

ORIGIN: ENDF/B-IV

WEIGHTING SPECTRUM:

NEDT1/401

FORMAT: Nuclear element data in SAM-CE format.

NUMBER OF GROUPS:

NUCLIDES: Data are included for 1H, 3H, 4He, 7Li, 10B, 11B, 12C, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, 182W, 183W, 184W, 186W, 235U, 238U, 239Pu, 240Pu.

ORIGIN: ENDF/B-IV

WEIGHTING SPECTRUM:

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ZZ-FEWG1: Coupled 37-neutron, 21-gamma group, P3, multigroup library in ANISN format, based on data from the DNA Working Cross Section Library and the Evaluated Nuclear Data File. Cross sections for the following materials are included: H, 3H, 6Li, 7Li, 9Be, 10B, 11B, 12C, N, O, F, Na, Mg, Al, Si, 31P, 32S, Cl, Ar, K, Ca, Ti, V, Cr, Mn, Fe, Ni, Cu, Mo, Ba, Gd, 181Ta, 182W, 183W, 184W, 186W,Pb, 235U, 238U, 238Pu, 240Pu. The energy range is 1.0E-11 to 19.64 MeV for neutrons and 0.01 to 14.0 MeV for gamma rays. 1/E and 300K Maxwellian weighting spectra were used for neutrons.

ZZ-GEDT1/400: Gamma-ray cross section library in SAM-CE format. Data are included for materials with atomic numbers 1-83, 86, 90, 92, 94, and for some naturally occuring elements. Data are given for the energy range from 1 keV to 100 MeV.

ZZ-GPDT1/402B: Gamma-ray production data in SAM-CE format. Data are included for 1H, N, O, Al, Si, Ca, Pb, 7Li, 10B, 12C, Mg, Fe, 235U, 238U, Cl, 182W, 183W, 184W, 186W.

ZZ-NEDT1/401: Nuclear element data in SAM-CE format. Data are included for 1H, 3H, 4He, 7Li, 10B, 11B, 12C, N, O, Na, Mg, Al, Si, Ca, Fe, Cu, Pb, 182W, 183W, 184W, 186W, 235U, 238U, 239Pu, 240Pu.

The retrieval program BCDEAN converts data from card image format to the binary form required by SAM-CE.

• N.M. Greene, et al.: "AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/B", ORNL-TM-3706 (March 1976).

• D. Garber, C. Dunford, and S. Pearlstein: "Data Formats and Procedures for the Evaluated Nuclear Data File, ENDF", BNL-NCS-50496 (ENDF 102) Revised Oct. 1975.

• D.E. Bartine, E.M. Oblow, F.R. Mynatt: "Radiation-Transport Cross-Section Sensitivity Analysis - A General Approach Illustrated for a Thermonuclear Source in Air", Nuc. Sci. Eng. 55, 147 (1974).

• W.W. Engle, Jr.: "A User'S Manual for ANISN, A One-Dimensional Discrete Ordinates Transport Code with Anisotropic Scattering", K-1693, Computing Technology Center (1976).

• J.J. Ritts, M. Solomito, P.N. Stevens: "The Calculation of Neutron-Induced Physical Doses in Human Tissues", ORNL-TM-2991 (1969).

• M.A. Abdou, C.W. Maynard, R.Q. Wright: "MACK: A Computer Program to Calculate Neutron Energy Release  Parameters (fluence-to-Kerma Factors) and Multigroup Neutron Reaction Cross Sections From Nuclear Data in ENDF Format", ORNL-TM-3994 (July 1973).

• "Protection Against Neutron Radiation", NCRP-38, National Council on Radiation Protection and Measurements    Washington, D.C. (1971).

• H.C. Claiborne, D.K. Trubey: "Dose Rates in a Slab Phantom from Monoenergetic Gamma Rays", Nucl. Appl. Tech. 8, 450 (1970).

• W.S. Snyder and C. Neufeld: Radiation Res., 6, 1, 67 (1957).

• B.J. Henderson: "Conversion of Neutron or Gamma Ray Flux to Absorbed Dose Rate", XDC 59-8-179 (August 14, 1959).

• E.A. Straker and M.L. Gritzner: "Neutron and Secondary Gamma-Ray Transport in Infinite Homogeneous Air", ORNL-TM-4464 (1969).

• J.V. Pace, III, D.E. Bartine, F.R. Mynatt: "Neutron and Secondary-Gamma-Ray Transport Calculations for 14-MeV and Fission Neutron Sources in Air-Over-Ground and Air-Over-Sea-water Geometries", ORNL-TM-4841 (August 1975).

• P.G. Fisher, W.S. Knapp: "Aids for the Study of Electromagnetic Blackout", 70TMP-12, General Electric - TEMP, Santa Barbara, Cal., p. 2.7 (July 1970).

D.E. Bartine, J.R. Knight, J.V. Pace, III and R.Roussin
Neutron Physics Division
Oak Ridge National Laboratory
Oak Ridge, TN, USA